ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 28

Document Title:

Site Profile for Allied Chemical 

Corporation Plant

Document Number: ORAUT-TKBS-0044

Revision: 00

Effective Date: 02/01/2006

Type of Document: TBD 

Supersedes: None

Subject Experts: Shelby L. Gubin and Robert Hysong 

Document Owner

Approval: Signature on File Approval Date: 01/31/2006

Cindy W. Bloom, TBD Team Leader

Approval: Signature on File Approval Date: 01/31/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/31/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/01/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 2 of 28

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/01/2006 00 New document to establish the technical basis for the development 

of a radiation exposure matrix for Allied Chemical Corporation Plant. 

First approved issue. Training required: As determined by the Task 

Manager. Initiated by Cindy W. Bloom.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 3 of 28

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations..................................................................................................................

PAGE

5

1.0 Introduction ................................................................................................................................. 6

2.0 Site Description and Operational History ....................................................................................6

 2.1 Site Description.............................................................................................................. 7

 2.2 Process Description ....................................................................................................... 9

 2.3 Source Term ................................................................................................................ 15

 2.4 Safety ........................................................................................................................... 17

 2.4.1 Visual Observation of Contamination ................................................................17

 2.4.2 Air Activity Measurement of Contamination ......................................................17

 2.4.3 Surface Contamination Measurements .............................................................18

 2.4.4 Decontamination ...............................................................................................18

 2.5 Incidents....................................................................................................................... 18

 2.6 Physical Examinations – X-Rays ................................................................................. 18

 2.7 Summary Operational Period Assumptions, Workdays, Work 

 Hours, Work Categories .............................................................................................. 19

 2.8 Residual Contamination............................................................................................... 19

3.0 Estimating Internal Exposure....................................................................................................19

 3.1 Uranium........................................................................................................................ 19

 3.1.1 Uranium Urinalysis Data....................................................................................19

 3.1.2 Uranium In Vivo Data ........................................................................................20

 3.2 Other Radionuclides..................................................................................................... 21

 3.3 Radon........................................................................................................................... 22

 3.4 Notations on Bioassay Records ................................................................................... 22

 3.5 Occupational Internal Dose Reconstruction Assumptions and 

 Summary ..................................................................................................................... 22

4.0 Estimating External Exposure...................................................................................................23

 4.1 Occupationally Required Medical X-Ray...................................................................... 24

 4.2 Miscellaneous Information about External Dose.......................................................... 24

 4.3 Occupational External Dose Reconstruction Assumptions and 

 Summary ..................................................................................................................... 24

5.0 Estimation of Exposure from Residual Radioactivity ................................................................24

References ...........................................................................................................................................27
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LIST OF TABLES

TABLE TITLE PAGE

2-1 Radionuclide activity ratios relative to uranium based on activities reported in ACCP 

concentrates, wastes, air and effluents.....................................................................................16

2-2 Source term activity ratios relative to uranium (based on the largest reported ratio relative to 

uranium)....................................................................................................................................16

3-1 Natural uranium’s specific activity and isotopic fractions ..........................................................19

3-2 Assumed types and ratios for intakes of other radionuclides relative to uranium

 intake activity ............................................................................................................................22

4-1 External exposure summary .....................................................................................................24

5-1 Residual radiation exposure summary......................................................................................26

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Plot Plan of Allied Chemical Corp. Plant in Metropolis, IL ..........................................................8

2-2 Process steps for conversion from ore concentrates (yellow cake) to UF6...............................10
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ACRONYMS AND ABBREVIATIONS

ACCP Allied Chemical Corporation Plant

ADU ammonium diuranate 

AEC U.S. Atomic Energy Commission

cc cubic centimeter

Ci curie 

DOE U.S. Department of Energy

dpm disintegrations per minute

ft foot

hr hour

keV kilovolt-electron, 1,000 electron volts

kg kilogram

L liter

m meter

MeV mega-electron volts, 1,000,000 electron volts

ml milliliter

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRC Nuclear Regulatory Commission

pCi picocurie

ppm parts per million

U.S.C. United States Code

WLM working level month

y year

α alpha particle

γ gamma

μCi microcurie 

μg microgram
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 6 of 28

1.0 INTRODUCTION

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist the National Institute for Occupational Safety and Health 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” (AWE facility) or a “Department of Energy facility” as defined in the Energy 

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. Sections 

7384I(5) and (12)].

EEOICPA, as amended, provides for employees who worked at an AWE facility during the contract 

period and/or during the residual contamination period. 

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when the 

AWE was processing or producing material that emitted radiation and was used in the production of 

an atomic weapon), or (2) during the residual contamination period (i.e., periods that NIOSH has 

determined there is the potential for significant residual contamination outside of the period in which 

weapons-related production occurred). For contract period employment, all radiation exposures must 

be included in dose reconstructions. For residual contamination period employment, only the 

radiation exposures defined in 42 U.S.C. § 7384n(c)(4) (i.e., radiation doses received from DOE/AECrelated work) must be included in dose reconstructions. This site profile covers only exposures 

resulting from nuclear weapons-related work. Exposures resulting from non-weapons related work, if 

applicable, will be covered elsewhere.

2.0 SITE DESCRIPTION AND OPERATIONAL HISTORY

The information that follows applies to a period of AEC operations at Allied Chemical Corporation 

Plant (ACCP) located in Metropolis, Illinois from January 1, 1959, to December 31, 1976, involving 

AEC-contracted conversion of uranium ore concentrates to uranium hexafluoride. Opening 

ceremonies for Allied Chemical's fluorine plant at Metropolis, Illinois occurred in October 1958 (Sloop 

1978). The original license issue date was December 17, 1958 (NRC 2003, p. 8-9). Four claims 

(numbers redacted) include bioassay results in December 1958, and one of these from December 15, 

1958, has a positive result indicating that uranium exposures might have occurred in late 1958 prior to 

the DOE Office of Worker Advocacy-established covered period. The plant was closed temporarily on 

June 30, 1964 and AEC (1966) reported that it might reopen in 1966. A review of available worker 

dosimetry records indicates that some workers might have been laid off or transferred around this 

period, but that other workers were still onsite. This analysis assumed that the residual contamination 

period was from January 1, 1977, through the present. 

Detailed information regarding buildings and processes is provided in Sections 2.1, 2.2 and 2.3 and is 

based on descriptions written after the period of operations covered by this document. Individuals, 

who worked at Allied Chemical Corp. Plant, indicate there was little variation in the processes over 

time. Because individual dosimetry results are available for the vast majority of workers at Allied 

Chemical, the reconstructed doses will be based on the specific individual dosimetry. To account for 

intakes of unmonitored radionuclides, bounding radionuclide fractions based on measurements in 

later years were used to estimate intake ratios of these radionuclides relative to uranium. Overtime, it 

is believed that the separations of uranium and associated radionuclides would have become more 
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efficient and that the radionuclide fractions derived from measurements in later years would 

overestimate exposures to non-uranium radionuclides in the earlier years. 

The ACCP radiological source term started with the receipt of uranium ore concentrates. Some longlived uranium progeny were included in the concentrates, e.g., thorium-230 and radium-226. Uranium 

chemical forms include oxides, fluorides and hexafluorides, which mean that exposures could have 

been to uranium solubility types F, M and/or S. 

Allied Chemical Corporation Plant was also known as General Chemical Division, Allied Signal 

Metropolis Plant and was later purchased by Honeywell.

2.1 SITE DESCRIPTION

“On February 4, 1957, the Allied Chemical and Dye Corp. announced selection of Metropolis Ill., as 

the site of its plant to process 5,000 tons of U3O8 a year under contract with the [Atomic Energy] 

Commission” (AEC 1957). The official plant startup date appears to be January 1, 1959, but there are 

indications that uranium and uranium workers might have been onsite in late 1958. One individual 

reported participating in opening ceremonies at Allied Chemical's fluorine plant at Metropolis, Illinois in 

October 1958 (Sloop 1978), indicating perhaps that uranium exposure could have occurred as early 

as October 1958. 

Allied Chemical operated its UF6 production facility using a “dry” conversion process (versus a “wet” 

solvent extraction process) to supply UF6 feed for the Paducah Gaseous Diffusion Plant through June 

30, 1964 under an Atomic Energy Commission contract (Perkins 1982). It was also noted that Allied 

Chemical was sampling uranium concentrates for other entities as of 1982.

On June 30, 1964, the plant was temporarily closed. AEC (1966) reported that the plant “may be 

reopened in 1966 for uranium hexafluoride production.” Perkins (1982) noted that Allied resumed 

operations in February 1968, but the available dosimetry records indicate that radiological exposures 

were still occurring at the plant between June 30, 1964 and February 1, 1968. Some of the claimant 

interviews and dosimetry records show that a number of workers were onsite during the shutdown, 

and hiring was occurring in 1967, although these records also indicate that a number of workers were 

laid off in July 1964. The available records do not indicate what activities might still have been 

ongoing in the plant during the shutdown period, although based on DOL job descriptions associated 

with one claim (number redacted), some operations associated with processing might have continued 

at least on a sporadic basis. 

Allied Chemical Corp.’s NRC license (NRC 2003) states, “The Allied Signal Metropolis Plant is located 

on approximately 1,000 acres of land in Massac County at the southern tip of Illinois, along the north 

bank of the Ohio River. The site perimeter is formed by US Highway 45 to the north, the Ohio River to 

the south, an industrial coal blending plant to the west and privately owned, developed land to the 

east. Plant operations are conducted in a single fenced-in, restricted area covering 59 acres in the 

north-central portion of the site.” The plant is now owned by Honeywell and still providing UF6 to 

gaseous diffusion plants for use in the commercial sector. The 1982 building locations are shown in 

Figure 2-1. A description of the buildings is excerpted from Allied Chemical Corp.’s licensing 

documentation (NRC 2003, pp. 8-8 to 8-9). The list was originally numbered from 8.5.1 through 8.5.8, 

but has been simplified to 1 through 8 below. 

Most of the uranium processing equipment is housed in a six-story structure termed the 

Feed Materials Building where essentially all of the steps in the UF6 manufacturing 

process are conducted. Other areas and buildings in which operations are conducted 
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involving the handling or processing of significant quantities of source material include 

the following:

1. A Sampling Plant which receives samples or concentrates for uranium assay 

and moisture content.

2. The Sodium Removal (pretreatment) and Uranium Recovery Facilities which 

are housed in buildings where high sodium content ore concentrates are 

treated to remove sodium impurity, and where materials are reprocessed to 

recover additional uranium.

3. The KOH muds washing facility which removes fluorides and KOH from the 

potassium diuranate muds generated in the fluorination scrubber system. The 

Figure 2-1. Plot plan of Allied Chemical Corp. Plant in Metropolis, IL (Perkins 1982, p. 3) 

washed potassium diuranate is then processed through Sodium Removal. The 

wash liquors are neutralized at EPF [environmental protection facility, this 

facility probably did not exist during the early years]. 

4. The Calcining Facility which dries the incoming feed material and recovered 

uranium as the first step in ore preparation.
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5. The Pond Mud Calciner Drier where hard/wet ore concentrates and KOH Muds 

are processed prior to packaging for blending with additional ore concentrates 

at the Feed Materials Building for conversion to UF6. 

6. The Laboratory Building which houses facilities for conducting process control, 

product, and radiological control analyses.

7. The Cylinder Wash Building where UF6 product cylinders are periodically 

washed and hydrostatically tested prior to reuse.

8. Outdoor pads for storage of drums of ore concentrates and other uranium 

bearing materials, as well as UF6 product cylinders.

Additional plant facilities which are involved directly in the UF6 manufacturing process 

but do not involve the handling of any significant quantities of source material include a 

fluorine manufacturing building, a fluoride waste treatment facility with four large CaF2

settling tanks, a powerhouse, a reductor off-gas incinerator, and two small uranium 

settling ponds to collect any uranium spills.

Allied was also involved in the manufacture of 30,000 pounds per week of liquid fluorine, 1200 tons 

per year of sulfur hexafluoride, 2500 pounds per week of antimony pentafluoride and 10,000 pounds 

per week of iodine pentafluoride (Perkins 1982). These materials were not radioactive.

2.2 PROCESS DESCRIPTION

The conversion process from ore concentrates (yellow cake) to UF6 is described below and a process 

flow diagram is presented in Figure 2-2. As of 1982, the processing capacity was 14,000 tons of 

uranium annually and the plant had processed over 100,000 tons of uranium from both domestic and 

foreign uranium concentrates (Perkins 1982).
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Figure 2-2. Process steps for conversion from ore concentrates 

(yellow cake) to UF6 (NRC 2003)

As part of an evaluation of environmental control technologies for nuclear fuel conversion (UF6) 

facilities, Perkins (1982) described the conversion of yellow cake to uranium hexafluoride at the Allied 

Chemical Corp. Plant. Although some changes in the process might have occurred since the first 

days of operation, it is likely that these changes would have been related to ore sampling techniques, 

ventilation/filtration controls and waste disposal practices, and that the general process was similar 

throughout the plant’s history. Process descriptions from Perkins (1982) follow. 

Introduction

The Allied facility uses the fluoride volatility process to produce UF6 from uranium 

concentrates. The facility does not accept uranium slurries.

When the 55-gallon-drummed concentrate is received it is weighed, sampled, and 

analyzed for moisture content. After feed preparation, mechanically sized particles 

(U3O8) are reacted with hot cracked ammonia to form uranium tetrafluoride (UF4) or 

green salt. Next, the green salt is combined at high temperature in a fluidized bed with 

gaseous fluorine to form the gas UF6. Finally, the UF6 off-gas is further purified by 

condensation and fractional distillation.

Each of these steps, with the resulting waste and effluent streams, is discussed in 

detail in the following sections.

Receiving and Sampling

Uranium concentrates are received at Allied Chemical Corp. Plant in 55-gallon drums. 

The drums are stored outside on specially constructed pads designed to allow 
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rainwater to drain to centralized collection sumps, which is then pumped to uranium 

spill control ponds. 

Drums are weighed, and a falling stream method is used to obtain a representative 

sample. The airborne dusts generated in the sampling process are removed through 

the use of two baghouses in series, which discharge through a stack. Empty drums 

are air cleaned and the discharge runs through two baghouses before being 

discharged to a stack.

Concentrates that are found to contain a high percentage of sodium are pretreated in 

the sodium-removal facility.

Sodium Removal (Pretreatment)

Because sodium forms a compound that causes caking and sintering in the fluorination 

fluid beds, any incoming uranium concentrates containing sodium and all the uranium 

concentrates produced by the uranium recovery sections of the UF6 facility must first 

have the sodium or potassium removed before they can be processed. These 

concentrates are treated by reaction with a solution of ammonium sulfate in four 

counter current vessels to make (NH4)2U2O7 for further processing. The chemical 

reaction is

Na2U2O7 + (NH4)2SO4 → (NH4)2U2O7 + Na2SO4

The liquid effluent contains the excess (NH4)2SO4, Na2SO4, K2SO4, and uranium and 

small quantities of 226Ra and 230Th, selenium, molybdenum, and other trace impurities. 

The effluent is sent to the two uranium spill control ponds. Uranium contaminated 

storm water is also discharged to these ponds. The sludges are allowed to settle, and 

the decantate is discharged to the main discharge outfall for the plant.... When the 

minimum freeboard is reached (approximately 2 ft) on a pond, the pond is dredged and 

the sludge is sent to the pond sludge calciner.... Pond liners are inspected and 

repaired at this time. In addition, all pond liners are underlain by a gravel layer that 

allows any seepage to drain to a leak detection sump.... 

The (NH4)2U2O7 (ADU) [ammonium diuranate] is removed from solution in the last 

reaction vessel and is sent to the main feed preparation section of the plant. 

Feed Preparation

In the feed preparation circuit, sodium-free uranium concentrate, either from the 

pretreatment section or from the sampling section, is first fed into a calciner where 

water and ammonia are removed. The off-gases containing ammonium, sulfur dioxide, 

uranium, and contaminant particles pass through two baghouses in series. 

Particulates recovered by blowback of the baghouses are collected and sent back into 

the blending feed stream.... 

The calcined material is blended and then agglomerated, dried, crushed, and sized 

before being sent to the reduction circuit. Wet off-gas streams join the off-gases from 

the calciner, whereas dried off-gases pass through two baghouses in series before 

discharge.... Material collected by the dry stream baghouse is fed into the blending 

feed stream. 
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Reduction

In the reduction circuit, Allied operates two trains in parallel, both having the same 

basic design. The U3O8 from the feed preparation circuit is fed into the reductor vessel 

where hot cracked ammonia (N2 and H2) and additional nitrogen mix with the feed to 

form a fluidized bed. The U3O8 is reduced to UO2, which is withdrawn from the bottom 

of the bed. Off-gases from the reactor include H2, N2, H2S, AsH3, vaporized S, SeH2, 

and particulates composed of UO2, unreacted U3O8 and reduced compounds of 

impurities originally in the concentrate.... 

The off-gases pass through two porous metal filters in series, a sulfur condenser, and 

incinerator.... 

Particulates collected by cleaning the filters are sent to the uranium recovery section.

Off-gases from the oxide vacuum cleaner are sent through cyclone and two baghouses 

in series before discharge.... Residues from these collection devices are sent to the 

uranium recovery section.

Hydrofluorination

Again in the hydrofluorination circuit, two trains in parallel are in use. The UO2 from 

reduction is fed into two reactor vessels in series, which use vaporized hydrofluoric 

acid and N2 to fluidize the UO2 and permit HF to react with the material to form UF4. 

The off-gases contain the excess HF needed to give good conversion to UF4 (10% 

excess or more), volatile SiF4, BF3, and some of the molybdenum and vanadium (which 

were present as impurities in the concentrate) as volatile fluorides and oxyfluorides, 

and any remaining sulfur as H2S4. These gases first pass through two sets of porous 

carbon filters per set, where the unreacted UO2 particles, entrained UF4 particles, and 

any other particulate matter are removed. The gases then pass through two venturi 

water scrubbers where the H2O in the off-gases condenses, and finally through a 

venturi KOH scrubber and a packed tower using KOH scrubbing liquid.

The residue from the clean carbon filters is sent to uranium recovery. Liquid coming 

from the venturi water scrubbers contains hydrofluoric acid and is sent to the acid 

neutralization treatment plant, where lime is added to precipitate the fluorine as CaF2. 

The uranium content of this stream is reported to average less than 5 ppm. Silicon, 

boron, and some molybdenum, and vanadium, which were originally in the 

concentrate, are expected to be in the sludge. The liquid from the KOH venturi and 

packed tower is sent to the liquid treatment systems.

The vacuum cleaning off-gases from the cleaning activities necessary in the 

hydrofluorination section are routed through a cyclone and two baghouses before 

discharge.... Residues from the dust collectors are sent to the uranium recovery 

section.

Fluorination

The green salt, UF4, produced in the hydrofluorination process is fed into one of two 

fluid bed reactors used in parallel. Again, Allied uses two trains at the works. The bed 

material is CaF2, and unreacted UF4. The vessels are operated at temperatures of 
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approximately 1000° F, and good temperature control is required. Fluorine, F2, from an 

on-site fluorine production facility is introduced into the reactor vessels to convert UF4

into gaseous UF6. 

The off-gasses, including uranium UF6, F2, HF, unreacted UF4, and other particulates, 

and volatilized impurities...are first cooled before passing through two sets of sintered 

nickel filters, each containing two filters. The material recovered as the filters are 

cleaned is stored, and then sent to the uranium recovery section. 

The stream exiting from the filters has the UF6 removed by condensation in a set of 

three cold traps used in series. Any uncondensed UF6, F2, HF and other volatiles 

passing from the cold wraps are removed from the gas stream by a KOH spray tower, 

a KOH pack tower, and finally a KOH coke box.

Carbon dioxide in the spent scrubber liquor reacts with the uranium and potassium as 

follows.

2(UO3)

4 - + 6CO2 + 6OH- + 2K+ → K2U2O7 + 6CO3

2- +3O2 +3H2O 

The K2U2O7 is settled from the KOH solution and sent to the pretreatment facility. 

Decantate from the uranium settling section is sent to the KOH regeneration system as 

indicated previously.... The KOH is then sent back to the scrubbing system. In 

addition to CaF2 compounds of uranium, vanadium, and molybdenum, trace quantities 

of compound of silica, carbon sulfur, and other trace contaminants are expected to be 

in the sludge.... 

The condensed material in the three cold traps in series is melted and drained to the 

still feed tanks. Entrained hydrogen fluoride is vaporized during melting and passes to 

the off-gas system. 

From the still feed tanks, the liquid is fed to a low-temperature boiler bubble cap 

column in which the impurities are volatilized and exit from the top of the column.... 

The liquid uranium hexafluoride passes from the low boiler column into the bubble cap, 

high boiler column. In this column, the UF6 is volatilized and exits from the top of the 

column. Non-volatilized impurities are removed from the bottom of the column and 

stored. 

The gaseous UF6 from the column is condensed into two cold traps operated in series. 

The UF6 is transferred to the shipping cylinders by melting the UF6 and allowing the 

material to drain into the cylinders.

Because impurities build up on the recycled CaF2 bed material used in the fluorination 

reactors, part of the bed material must be withdrawn periodically. This “ash” [also 

referred to as spar or bed material] along with the fluorination-cleaned filter residue, is 

drummed and stored for a minimum of 6 months to permit the uranium daughters 234Th 

and 234Pa to decay. The ash is then sent to the uranium recovery section. A total 

quantity of about 0.1 ton of ash per ton of uranium processed is produced, and it 

contains approximately 1.8% of the total uranium processed and most of the uranium 

daughters originally present in the concentrate.
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The dusts produced by withdrawing bed material from the reactors are passed through 

a cyclone and two baghouses in series before discharge.... The ash vacuum cleaner 

systems also discharge from this stack. Residue from cleaning these particulate 

removal devices is treated similarly to the spent bed material.... 

The feed preparation, reduction, hydrofluorination, and fluorination circuits are all 

located in the feed material building. This building has a complete air changeout

approximately once every 5 minutes.... 

Washdown of the feed materials building is pumped from collection sumps into the 

uranium recovery ponds.

The transfer lock to fluorination discharges through two baghouses.... Again, dust 

collection residue is sent to uranium recovery. 

Uranium Recovery

There are several sources of uranium-containing wastes produced by the facility, and it 

is desirable to recover the uranium from these wastes. Thus, uranium-containing dried 

sludges, spent bed material, filter residue, and various scraps and dust are sent to the 

uranium recovery circuit. 

Off-gases from storage before processing vent through the 3-1 stack. The main 

contaminants in the stack off-gases should be radon and radon daughters, if the 

radium originally in the concentrate feed goes into the ash (as was assumed in the 

previous section).

The material from storage is first dumped and ground. Dusts from this operation are 

recovered in a baghouse.... 

The finely ground material is leached using a sodium carbonate solution to solubilize 

the uranium as the tricarbonate complex.... 

After leaching, the pregnant solution is filtered. Lime is used to aid in precipitation of 

any fluoride, and the solution is again filtered. Sludge from the filters is dried. Offgases from the dryer pass through the dumping and grinding off–gas baghouse.

The dried sludge probably contains all nondecayed uranium daughters, some residual 

uranium, and some fluoride compounds. This sludge is drummed and sent to a 

licensed radioactive waste disposal facility. Approximately 1500 tons of this waste, 

containing a maximum of 46.9 Ci is produced yearly. 

The uranium is precipitated from the solution using NaOH. The uranium is filtered and 

washed and then sent to the sodium removal section.

Cylinder Wash

Cylinders are returned to Allied from the enrichment facilities. These cylinders contain 

residual UF6 and the daughters of uranium that have “grown in” while the cylinders 

containing UF6 were in storage. The daughters in general are found plated out on the 

cylinder walls.
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After any residual UF6 is removed from the cylinder, the cylinder is washed, using a 

solution of Na2Co3 to remove any impurities from the walls and to solubilize the 

uranium. The wash solution is filtered to remove the unleached solids, and the 

pregnant solution is pumped to join the pregnant solution in the uranium treatment 

section.

The solid residue from the filters contains daughter products of uranium, principally 

234Th and 234Pa, and is stored on-site in drums until disposal in a licensed waste 

disposal facility.

Sludge Dryer

As previously described, the process generates several sludges that contain uranium. 

Before being sent for uranium recovery, these sludges are dried in a calciner. Offgases containing SO2, HF and particulates from the calcining operation pass through a 

baghouse and a water spray tower before discharge.... 

The spent scrubber water, which contains uranium, fluoride compound, and some 

sulfur compounds, is sent to the uranium recovery ponds.

The calcined sludges are drummed and stored until they can be processed in the 

uranium recovery section.

2.3 SOURCE TERM

The plant produced UF6 from uranium ore concentrates initially for gaseous diffusion facilities and 

later also for commercial fuel fabrication facilities. The uranium conversion process at the plant has 

remained essentially unchanged since its inception. In 1982, Perkins (1982) reported that the 

concentrates initially contained 70 to 75% uranium by mass, and these concentrates included 

impurities that were dependent on the mines and mills where the yellow cake was derived. 

Uranium chemical forms at Allied Chemical Corp. Plant included all solubility types, i.e., F, M and S. 

The initial design processing capacity was 5,000 tons of uranium in a year, and by 1982 the 

throughput capacity had been increased to 14,000 tons of uranium in a year in what was essentially a 

continuous chemical process. As of 1982, over 100,000 tons of uranium had been processed 

(Perkins 1982). 

The initial milling of the uranium prior to arrival at Allied Chemical Corp. Plant resulted in 

disequilibrium of uranium progeny in the concentrate. Exposure to the non-uranium radionuclides 

occurs at the front end of the processing, beginning with receipt. Increased concentrations of other 

radionuclides relative to uranium are found in the fluorination bed ash and other solid waste streams, 

such as the sludge. Environmental air measurements in the late 1970s showed ratios of Ra-226 and 

Th-230 relative to uranium that were larger than the concentrate ratios, but lower than complete 

equilibrium ratios.

Radon concentrations would probably have been elevated during the initial opening of the drums of 

concentrate, in the feed area and in work areas where radium would have concentrated. No 

measurements of radon concentrations have been found at this time. 

Perkins (1982) summarizes uranium, thorium and radium-226 activities in Allied Chemical effluents 

and wastes from the 1970s. These values, shown in Table 2-1, are used to estimate ratios of the 
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activity of other radionuclides to uranium in the effluent. The reported concentrate ratio was based on 

an average of yellowcake feed ratios from 33 mills (Perkins 1982, Table II-8). The equilibrium activity 

ratio of Th-230 to uranium in uranium ore would be about 0.4886; it can be seen that the milling of the 

ore usually reduces this ratio by almost a factor of 100, by removing the Th-230 from the concentrate. 

It should also be noted that Th-232 comprises a portion of the radioactivity in the uranium 

concentrates, and although Th-232 wasn’t specifically reported for most effluents at Allied, the water 

effluents for 1975 and 1976 show annual average Th-232:Th-230 activity ratios of 0.076 and 0.017, 

respectively. 

Table 2-1. Radionuclide activity ratios relative to uranium based on activities reported in ACCP 

concentrates, wastes, air and effluents. 

Source

Average Largest ratio

Th-nat:U-nat Th-230:U-nat Ra-226:U-nat Th-nat:U-nat Th-230:U-nat Ra-226:U-nat

Concentrate (yellow 

cake feed averaged 

over 33 mills)

5.20E-03

Uranium recovery solid 

waste (dried sludge)

3.58E+00 1.90E+01 1.11E+00

Solid nonsludge waste 5.27E-03 2.75E-04

Outdoor air 6.70E-02 1.97E-03 1.67E-01 8.46E-03

Air effluent 1.70E-02 2.70E-04 4.75E-02 3.38E-04

Water effluent 3.99E-03 2.04E-02 1.58E-02 5.79E-02

Processing at Allied Chemical Corp. Plant results in further separation of uranium and progeny, and 

this can result in increased activity ratios of non-uranium radionuclides in some areas in the plant. 

The largest disequilibrium ratios are associated with the solid waste (dried sludge) from the uranium 

recovery facility and these are used to derive a source term that would result in the largest internal 

dose. In reality, it is unlikely that any one worker would be consistently exposed to these 

disequilibrium ratios for an entire career. It is assumed that the activity of Th-228 is in equilibrium with 

Th-232. To derive the source term in Table 2-2, it is assumed that the short-lived progeny of internal 

dosimetric interest are in equilibrium with their closest long-lived parent. 

Table 2-2. Source term activity ratios relative to uranium (based on 

the largest reported ratio relative to uranium in Table 2-1). 

U-natural 1

Th-232 3.58

Th-230 19.0

Th-228 3.58

Ra-226 1.11

Pb-210 1.11

Po-210 1.11

Pa-231 0.887

Ac-227 0.887

AEC and U.S. Department of Energy (DOE) inventory reports use a designation of JSA to identify 

Allied Chemical Corp. Plant. DOE (Dalton 2000) reports that the Rocky Flats Plant shipped 71 kg 

(160 pounds) of depleted recycled uranium to Allied in 1975, which is a very small fraction of the 

annual inventory (which might have been as high as 10 to 28 million pounds) processed at Allied. A 

second mention of the same shipment indicates that the 71 kg were shipped to Allied Chemical Corp. 

in Idaho Falls, Idaho. Another reference (Author unknown c. 2000) referred to a shipment of 22,060 

kg from Rocky Flats to Allied Chemical Corp. between 1955 and 1964, but this appears to be an error 

due to a mix up between the next designation in the report, LAB, with JSA. It is assumed that 

contaminants in recycled uranium were not significant contributors to Allied workers’ radiation dose.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 17 of 28

Allied is licensed by the NRC to have sealed Cs-137 sources up to 100 millicuries each. These are 

used as level control devices and are unlikely to contribute to internal dose.

2.4 SAFETY 

Specific radiological safety measures implemented in the beginning of Allied’s operations are not 

available, but as of 1958, radiological control programs in general had become more sophisticated. 

Dosimetry is available from the earliest days of covered operations. The following information is 

based primarily on reference material dated after 1972, and it should be noted that the criteria to 

prompt radiological control activities could have differed in the early years. 

Engineered confinement was and is the primary source of contamination control used at Allied 

Chemical. Drums, process vessels, lines, etc. are used to confine the materials. When unconfined 

uranium is handled, procedures call for ventilation controls to reduce exposures. (Allied c. 1985, p. 

27). 

An exception was requested from the NRC’s requirement to post and label individual radioactive 

material areas and containers, and is replaced by worker training (NRC 2003, p. 11-1). Signs state 

that any area or container in the plant may contain radioactive materials.

Protective clothing and shoes or shoe covers were provided to employees and visitors to Allied 

Chemical to assure that personnel were not inadvertently contaminated with uranium compounds as 

of 1982 (Perkins 1982), and probably from the earliest days of operations. This did not apply to 

administrative areas where contamination would not be expected.

Current Allied procedures, discussed below, include four contamination measurement and control 

parameters:

• Visual observation of contamination

• Air concentration measurements

• Surface contamination measurements 

• Decontamination procedures 

2.4.1

Allied noted that natural uranium compounds have a low specific activity, characterized by a high ratio 

of mass to activity. The Allied Chemical uranium compounds are highly colored, which provides an 

immediate method for detecting gross surface and air contamination. For instance yellow cake was 

yellow and UF6 releases appear as white smoke due to the decomposition products, UO2F2 and HF. 

Procedures from 1985 stated that when employees detect small spills, they are responsible for 

cleanup. Larger leaks or spills that resulted in visible accumulation of uranium on equipment, or on 

the floor, required immediate decontamination, activation of the respirator lights, and possibly even 

shutdown of the unit. The foremen are to prepare a health physics incident report and notify the 

health physics shift technician to initiate special air sampling and investigation as appropriate. Health 

physics personnel make a daily inspection of uranium operating areas to ensure proper 

decontamination and contamination control. 

Visual Observation of Contamination

2.4.2

As of 1985, stationary breathing zone samples were collected daily from 57 stations in the uranium 

hexafluoride building, 10 in the sampling plant, 2 in the drum dumper, 3 in sodium removal and 2 in 

Air Activity Measurement of Contamination
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 18 of 28

the uranium recovery unit. Summaries of some of these measurements can be found in the available 

references. As of 1985, wearing of respirators was required when the average gross alpha air 

concentration on a floor exceeded 4 E-11 μCi/cc gross or when any 4 air samples on the floor 

exceeded that concentration. In addition, as of 1985, any single air sample result which exceeded 1 

E-10 μCi/cc required investigation and preparation of a health physics incident report (Allied c. 1985).

Although some air concentration data are available for the 1970s and later, these data are not used to 

estimate workers’ intakes because of the abundance of bioassay data.

2.4.3

Surface contamination measurements in uranium and non-uranium areas are made routinely. 

Removable contamination is determined by taking a wipe or smear of 100 cm2 on a surface. Each 

smear is counted for alpha radioactivity and reported in dpm/100 cm2 removable alpha contamination. 

The plant lunch room and offices or control rooms are surveyed weekly as are the locker rooms and 

personnel exit areas. Uranium processing areas such as the Feed Materials Building, Sodium

Removal, Uranium-Recovery Building, and Sampling Plant are surveyed monthly. Other non-uranium 

processing areas are surveyed quarterly.

Surface Contamination Measurements

As of 1985, decontamination is to be initiated when removable alpha contamination exceeds 5000 

dpm/100 cm2 in the uranium processing areas. All other areas are to be decontaminated when 

removable alpha contamination exceeds 200 dpm/100 cm2. 

One summary of Health Physics data (Wilkins 1992) for the last half of 1991, noted that 1.7% of the 

2,002 weekly smears exceeded the weekly limit of 200 dpm/100 cm2

, and the highest result was 923 

dpm/100 cm2 in the lunch room on a table. One of the 570 monthly smears and none of the 163 

quarterly smears exceeded the limit. 

2.4.4

Allied Chemical has detailed procedures for decontaminating spills or leaks found in the plant. 

Generally these procedures include roping off the area, sweeping up or vacuuming up the 

contaminated material and then washing off the contaminated surface. During these procedures 

protective clothing are to be used and monitoring is to be performed.

Decontamination

2.5 INCIDENTS

The claims include some information regarding incidents, but details are few and there are no dates or 

references to particular incident reports. Occasional leaking of valves or inadequate packing resulted 

in material releases. 

2.6 PHYSICAL EXAMINATIONS – X-RAYS

No documented Allied Chemical requirement for occupationally required medical x-ray examinations 

has been found. More than half of the claimant interviews indicate that chest x-ray examinations were 

routine and a number of the interviews indicate the frequency was annual. The examinations were 

offered annually, but were optional by 1980, and would have been done at Massac Memorial Hospital 

or perhaps a local doctor’s office according to one employee. 
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 19 of 28

2.7 SUMMARY OPERATIONAL PERIOD ASSUMPTIONS, WORKDAYS, WORK HOURS, 

WORK CATEGORIES 

Because bioassay data and film badge data were used to estimate internal and external exposures 

during the AEC uranium operations, an estimate of the number of workdays or work-hours per year is 

not important. 

No attempt has been made to sort workers into exposure categories. Individual dosimetry appears to 

be available for most workers. 

2.8 RESIDUAL CONTAMINATION 

Uranium concentrate processing continued at Allied Chemical Corp. Plant after 1976, and 

differentiation between residual contamination from the AWE period and contamination due to later 

and/or ongoing operations is not feasible. 

3.0 ESTIMATING INTERNAL EXPOSURE 

The primary source of internal radiation exposure at Allied Chemical Corp. Plant was uranium dust 

produced from the processing of uranium concentrates to produce uranium hexafluoride. It is 

assumed that the uranium was of natural enrichment prior to 1977, although there is some indication 

that Allied’s source term included a small amount of depleted uranium. The uranium ore concentrate 

also contained uranium progeny of dosimetric interest as well as thorium (assumed to be Th-232 and 

Th-228).

3.1 URANIUM

During the processing of uranium concentrates at Allied, workers might have been exposed to a 

variety of uranium chemical forms that encompass all lung absorption types: F, M and S. Although 

some process steps might have had more or less exposure to a given lung absorption type, it is not 

clear how well separated these areas were. In addition, workers may have worked in multiple 

uranium process areas. The selection of absorption type should depend on the organ of interest. The 

specific activity of the natural uranium and the isotopic mass and activity fractions are shown in Table 

3-1. For organ dose estimates, the uranium activity can be assumed to be from U-234. A conversion 

factor is also provided to adjust results reported as μg of U-235 to total uranium activity in pCi.

Table 3-1. Natural uranium’s specific activity and isotopic fractions.

Activity 

Fraction

Mass 

Fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.4886 5.37E-05 21.4 0.00745

U-235 0.0228 7.20E-03 1 1

U-238 0.4886 9.93E-01 21.4 138

Useful factors

U-natural pCi/ug 0.683

U (pCi): U-235 (ug) 94.8

 

3.1.1

Individual uranium urinalysis data began as early as December 1958 and was done in house. It 

appears that most workers will have bioassay data. The claimant file records show that bioassays 

Uranium Urinalysis Data
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 20 of 28

were typically collected at hire and periodically thereafter. Allied notes that fluorimetry was used for 

analyses until 2002, when kinetic phosphorescence analysis (KPA) began. Urinalysis data for the 

period 1989 to 1992 are missing, and Allied reported that it was assumed that the missing records 

might have been water damaged beyond recovery.

In the early years, uranium urinalysis results reported in units of microcuries per milliliter (μCi/ml) are 

misstated, and the units are actually in units of micrograms per liter (μg/L). The recording limit of 10 

μg/L is found in the early records and is consistent with the default detection threshold for uranium 

urinalysis assumed to be 10 μg/L based on a reported sensitivity of 5 to 10 μg/L for uranium 

fluorimetry urinalysis in the early years (Wilson 1958). Beginning in late 1974, a recording limit of 5 

μg/L appears to have been in use. Allied reported that their MDL was 2 μg/L in the early 1980s, but 

no records of MDL are available prior to that. In 1985, Allied Chemical Corp. Plant noted that routine 

urinalysis samples were to be left following 24 to 96 hours of absence from the workplace, but it is 

unknown when this procedure was first implemented. Also, as of 1985, Allied employees were being 

encouraged to leave special urine samples prior to their next work shift in response to known or 

suspected airborne exposures (NRC c. 1993, pdf p. 50). In later years, it was noted that a result 

equal to 30 μg/L was cause for investigation.

On the computer printouts, sometimes results are noted in the comments section rather than the 

bioassay measurement section. 

3.1.2

Lung counts are available for many workers beginning in late 1978, which is after the covered 

operational period. Lung counts appear to have been performed on an annual or biannual frequency 

until the mid-1980s. They appear to be less frequent in later years, but this may be that only some 

workers were required to have annual lung counts in later years. 

Uranium In Vivo Data

Allied claims include a letter that reports use of a uranium lung counting system from Radiation 

Management Corporation (RMC) and analysis of the spectra using a software package called 

RMCWBC, based on the Oak Ridge National Laboratory program, ALPHA-M. The counting system 

includes an RMC-designed shadow shield and four detectors, which appear to consist of thin NaI(Tl), 

and thicker CsI(Na) crystals. The dimensions are reported as 5 feet in diameter, but are probably 5 

inches in diameter. Prior to analysis of the spectra, the background is automatically subtracted. A 

REMCAL standard man equivalent phantom was used to calibrate the system. Radionuclides include 

U-235, Th-234, Cs-137 and K-40.

Specific Allied criteria are given for judging the data and can be used for considering the relative 

validity of a result: 

• Fit should be less than 3.0 and usually ranges between 0.5 and 1.5. An ideal fit would be 0.

• The shift value should range between +2.0 and -2.0, values outside the range of +/- 3.0 are 

carefully reviewed before they are accepted. MCA zero level and/or gain adjustments are 

warranted when the shift value exceeds +/-2.0.

• The gain should ideally be 1.0 and a gain of 0.90 to 1.10 is considered acceptable. 

Adjustment to the gain and/or zero may be need if the value is outside this range.

The analysis program reportedly prints a table showing all radionuclides found and their estimated 

activities and relative two standard deviation counting error in terms of a percent error. Allied reports 
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 21 of 28

that results of U-235 are given in μg and results for other radionuclides are given in nanocuries; 

however, some records show Th-234 results in μg. Results with errors greater than 100% (results 

less than two times their counting standard deviation) are considered questionable. The reported U235 minimum detection level was noted as 63 μg, which was equated to 32% of the maximum 

permissible lung burden. The maximum permissible lung burden was noted as 194.3 μg of U-235 or 

summed as 26,974 μg of total uranium (Allied 1985). In later years, it was noted that a result equal to 

50% of the maximum permissible lung burden was cause for investigation.

Unless there is a prominent low energy peak or detector ratios vary by more than 0.025, the detected 

uranium is assumed to be in the lung.

In general, the urinalysis results should be used to estimate intakes and the lung counts can be used 

as supporting data. The lung counts do not directly measure total uranium and although height and 

weight are recorded and could have been used to estimate chest wall thickness, there is no indication 

that chest wall thicknesses were measured directly.

To convert U-235 mass to total uranium activity, it is assumed that the uranium is of natural 

enrichment, and so the U-235 mass in μg is multiplied by 2.16 pCi/μg (the specific activity of U-235) 

and divided by 0.0228 (the activity fraction of U-235), a total factor of 94.8, to obtain the total uranium 

activity in pCi.

To convert Th-234 to total uranium activity, it is assumed that the Th-234 is in equilibrium with both U238 and U-234, and so the mass of Th-234 in μg is multiplied by 2.32E10 pCi/μg and divided by 

0.4886 to obtain the total uranium activity in pCi. If the Th-234 is given in nCi, it can be converted to 

pCi of uranium by dividing by 4.886E-4. Note that with a 24.1-day half-life, Th-234 might not be in 

equilibrium with its uranium parent depending on when and where an exposure occurred.

The K-40 and Cs-137 reported in lung counts are not likely to be associated with workplace 

exposures, and can usually be ignored for the purpose of dose reconstruction. 

A handwritten note (Honeywell c. 1998, pdf page 2) indicates that other whole body counting systems 

might have been in use as of 1996, but there were no differences in reported sensitivity. 

3.2 OTHER RADIONUCLIDES

Bioassays for non-uranium radionuclides do not appear to be in the claim files. To account for intakes 

of other radionuclides (except radon) handled incidentally to the uranium processing, the ratios 

relative to uranium activity in Table 3-2 can be used to estimate upper bound intakes relative to 

uranium. These ratios are reported for process streams where non-uranium radionuclides are 

concentrated. It is unlikely that any worker was exposed to this source term for their entire Allied 

Chemical Corp. Plant career, but these numbers represent a bounding estimate of intakes relative to 

uranium.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 22 of 28

Table 3-2. Assumed types and ratios for intakes 

of other radionuclides relative to uranium intake 

activity. 

Radionuclide Type

Activity ratio relative 

to uranium

Th-232 M, S 3.58

Th-230 M, S 19.0

Th-228 M, S 3.58

Ra-226 M 1.11

Pb-210 F 1.11

Po-210 M 1.11

Pa-231 M 0.887

Ac-227 F, M 0.887

3.3 RADON

No radon air concentration data have been found to date. Radon exposures were probably most 

significant in the receiving, sampling and feed areas and in the waste storage area, where radium 

source terms would have been more significant. It can be reasonably inferred that the radon 

concentrations would not have been higher than the high, but not atypical, concentrations during 

processing at Linde and Mallinckrodt, where African ores were being wet-processed in the 1940s and 

1950s. A concentration of 100 pCi/L, with a fraction of equilibrium of 0.4 would equate to 4.8 working 

level months per year (WLM/y).

3.4 NOTATIONS ON BIOASSAY RECORDS

Some of the bioassay records include notations and abbreviations. These include:

• Laid-off

• Rehired

• Trnfd [transferred]

• MTO [Morristown Office, corporate headquarters in N.J.]

• MPLB [maximum permissible lung burden]

• Sample type R [assumed to mean routine]

• Sample type S [assumed to mean special]

3.5 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The assumed operational exposure period ran from January 1, 1959, to December 31, 1976.

In general, the urinalysis results should be used to estimate uranium intakes and the lung counts can 

be used as supporting data. The uranium photofluorimetry urinalysis reporting limit in the early years 

at Allied Chemical Corp. Plant was 10 μg/L. Uranium may be either absorption type F, M or S. For 

dose calculations, uranium intakes can be assumed to be U-234. 

To estimate uranium intakes from the lung counts, the intake of U-235 is determined, and then 

adjusted to total uranium intake activity using the conversion in Table 3-1. The lung detection 

threshold is 63 μg for U-235 and 5970 μg for total uranium.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 23 of 28

Intakes of other radionuclides are accounted for by calculating the uranium intake(s) from the uranium 

urinalysis, and then applying the fractions in Table 3-2 to estimate intakes of other radionuclides. The 

dose distributions are assumed to be constant for the other radionuclides.

A high estimate (constant distribution) of radon exposure is 4.8 WLM/y.

Based on the current information, all workers were monitored, so at this time no assumptions have 

been provided for unmonitored workers or unmonitored periods. This section will be developed 

further, if necessary.. 

4.0 ESTIMATING EXTERNAL EXPOSURE

External dosimetry results are available for Allied Chemical Workers beginning January 1, 1959, and 

include results for gamma, beta and neutron exposures. Allied Chemical reports that the data prior to 

1970 is handwritten and the records do no include the company or method of measurement. 

Beginning in 1969, R.S. Landauer Jr. & Company provided film badge dosimetry services and 

measured gamma and beta radiation. In about April 1976, Eberline Instrument Company provided 

TLD dosimetry services. Neither Landauer nor Eberline reported doses below 10 mrem.

Film badge detection thresholds in the early days were assumed to be as high as 50 mrem for gamma 

and 60 to 80 mrem for beta. A detection threshold of 40 mrem has been recommended for this 

project (ORAUT 2004). 

The majority of photons from natural uranium have energies in the range of 30 to 250 keV, so this 

analysis assumed workers were exposed to photon energies in the 30 to 250 keV range, which is 

claimant favorable. Nonpenetrating dose from natural uranium consists primarily of electrons with 

energies above 15 keV. Because the uranium is mixed with material that can produce neutrons from 

an alpha particle collision, the neutron dose is also considered. For consistent presentation, exposure 

or dose is reported as: 

• Penetrating, assumed to be associated with photons of energies 30 keV or greater, and 

• Nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with 

electrons. 

• Neutron, assumed to be associated with neutrons of energies of 0.1 to 2 MeV.

The earliest available Allied Chemical Corp. Plant external dosimetry results are dated January 1, 

1959. Badge exchange frequencies at Allied were typically monthly in the available records. In the 

1980s, hourly employees exchanged badges monthly and salaried employees changed badges 

quarterly.

Although some records have results recorded as neutron doses, in later years no neutron doses are 

available. Most of the results appear to be zeroes and it is not clear if these represent an actual 

dosimeter result or are just written because there was no result. The Bridgeport Brass site profile 

(ORAUT 2005) notes that the minimum detectable neutron dose reported in the early 1960s was 60 

mrem. 

A quick and probably high estimate of neutron dose is estimated based on a 500-pound drum 

containing 75% by mass of natural uranium. It is assumed that the alpha particles collide with a 

fluorine target. Using the method described in ORAUT 2005, the neutron dose calculated at one foot 
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 24 of 28

from a drum is 3.28E-04 rem/hr. A worker, who is one foot from such a drum for 2000 hours per year, 

would be exposed to a dose rate of 6.56E-01 rem/y. This document considers this estimated neutron 

dose rate to be an upper bound of the Allied neutron exposure rate. 

4.1 OCCUPATIONALLY REQUIRED MEDICAL X-RAY

No definitive information was found that indicated the type or frequency of occupationally required 

medical x-ray examinations, although the claimant interviews support an assumption of an annual 

chest x-ray examination. The type of X-ray should be based on current ORAU Team guidance. 

Organ doses can be obtained from the current revision of ORAUT-OTIB-0006, Technical Information 

Bulletin: Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAUT 

2003).

4.2 MISCELLANEOUS INFORMATION ABOUT EXTERNAL DOSE

This section includes external dose information that could be of value for specific dose 

reconstructions. This analysis did not consider such information generically because of its limited 

applicability or because of limited details.

The fluorination bed ash contains short-lived progeny, which result in elevated beta and gamma 

radiation fields in the vicinity of the ash. Workers handling this material or washing drums or cylinders 

could receive extremity doses that are significantly higher than those measured by dosimeters worn 

on the trunk of the body. 

4.3 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Table 4-1 summarizes assumption regarding occupational external doses during the covered period 

January 1, 1959 to December 31, 1976 at Allied Chemical Corp. Plant.

Table 4-1. External exposure summary.

Exposure mode Exposure type Exposure or dose rate Basis

Exposure time 

assumption Annual exposure

IREP 

distribution

Uranium processing 

Penetrating - - - Based on individual 

dosimetry

-

Non-penetrating - - - Based on individual 

dosimetry

-

Neutron 0.328 mrem/hr Calculation 2000 work- hr/y 6.56E-01 rem Constant

Medical X-ray Initial plus one exam per 

year 

- - See ORAUT-OTIB-0006, (ORAUT 

2003) 

- Use generic external dose reconstruction guidance, when summary information is unavailable.

5.0 ESTIMATION OF EXPOSURE FROM RESIDUAL RADIOACTIVITY

Allied Chemical Corp. Plant continued to operate after the AWE covered period ended on December 

31, 1976. NIOSH (2004) noted that a significant potential for residual contamination remained after 

1976. It is not feasible to estimate exposures to residual activity from workplace measurements 

during continued operations; however, it is reasonable to estimate internal intakes due to residual 

contamination after 1976 by assuming that up to one additional year of intake occurred during the 

residual period as shown below.

During operations with uranium, airborne contamination can be caused directly from operations with 

uranium bearing materials as well as by the resuspension of surface contamination. This airborne 
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 25 of 28

contamination can be removed by settling on floors or other surfaces, or by ventilation (both 

engineered and natural building ventilation). The surface contamination levels can be increased 

directly by radioactive material processing/handling or by the settling of airborne contamination. This 

level is decreased by various incidental means such as general housekeeping or tracking of 

contamination out of the area on equipment or personnel.

To estimate the airborne contamination from residual contamination, it was assumed that during 

uranium operations, these factors reached an equilibrium that caused the airborne concentration and 

surface contamination to remain at a constant level. Once the contamination-generating source, the 

uranium recovery operation, was removed, the other factors then caused the mass of uranium to 

decrease at some rate due to the remaining removal factors.

The analysis of residual intakes assumes that 1) uranium operations directly accounted for half the 

airborne concentration in the facility during operations, (i.e., half of the previously calculated airborne 

exposure is assumed to come from resuspended materials) and 2) all the remaining removal 

mechanisms eliminated only 1% of the uranium mass per day. With these assumptions in place, an 

equation can be written for the airborne concentration as a function of time.

C(t) = 0.5C(0) × e-0.01t

where: 

C(t) = Airborne activity at time t (pCi/m3

) 

t = time since uranium operations ended (days)

C(0) = Concentration during operations (dpm/m3

) 

0.01 = fraction of material removed per workday 

When this is multiplied by the breathing rate of 9.6 m3 per 8-hour workday, a daily intake is calculated. 

This equation can then be integrated from the end of uranium operations to infinity to determine the 

total amount of uranium that could be inhaled from residual contamination. This integration results in:

I = 9.6 × 0.5 x C(0) / 0.01

where: 

I = total intake from inhalation of residual activity (dpm)

9.6 = breathing rate (m3/8-h workday)

A total uranium inhalation intake of 480 times the air concentration was assumed to occur during the 

residual contamination period. This compares to an estimate of annual intake during the operational 

period of 2,400 m3

/y of air breathed times the initial concentration, which would result in an intake 5 

times larger. This indicates that an estimate of inhalation intake during the residual period will be 

accounted for by assuming one additional year of intake at the calculated intake rate prior to 1977 for 

the individual. This assumed inhalation intake from residual contamination is sufficiently large to 

account for dose from ingestion intake.

To estimate external exposure to residual activity, external dosimetry results that are available after 

1976 are used to provide an upper bound of the dose, and this is assumed to be a constant 

distribution. 

Neutron doses from residual contamination are assumed to be insignificant.
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 26 of 28

To summarize, the following information in Table 5-1 is used to account for dose during the residual

exposure period.

Table 5-1. Residual radiation exposure summary.

Source Exposure information Period of applicability Annual exposure

IREP 

distribution

Uranium and progeny 

contamination 

- Beginning January 1, 1977 Add up to one additional year 

of inhalation intake based on 

individually calculated results

Constant

Uranium contamination Penetrating (Photons, 30-250 

keV, AP, acute)

Beginning January 1, 1977 Use reported dosimetry results Constant

Uranium contamination Nonpenetrating

(Electrons >15 keV, acute)

Beginning January 1, 1977 Use reported dosimetry results Constant
Document No. ORAUT-TKBS-0044 Revision No. 00 Effective Date: 02/01/2006 Page 27 of 28

REFERENCES 

Author unknown, c. 2000, page of summary information regarding Allied Chemical Corp. Plant, [p. 27 

of Allied Chemical Corp., IL.pdf].

AEC (U. S. Atomic Energy Commission), 1957, excerpt from Major Activities in the Atomic Energy 

Programs, January-June 1957, United States Government Printing Office, January, p. 9 [pdf p. 

13-14 of 27, Allied Chemical Corp., IL.pdf]. 

AEC (U. S. Atomic Energy Commission), 1966, excerpt from Annual Report to Congress of the Atomic 

Energy Commission for 1965, United States Government Printing Office, July, p. 37 [pdf p. 17-

18 of 27, Allied Chemical Corp., IL.pdf].

Allied Chemical Corporation Plant, c. 1985, excerpts from Site Policies and Procedures, 1985 

[030000439.pdf]. 

Dalton, H. F., 2000, memorandum plus attachment to J. J. Fiore (Deputy Assistant Secretary), Office 

of Site Closure, “Recycled Uranium Project – Final Site Report Rocky Flats Environmental 

Technology Site,” Colorado, June 26 [pdf pages 74 and 115 of Rocky Flats\04-27-05 - Data 

Capture\030003364.pdf].

Honeywell, “Current Site Procedures,” circa 1998 [030000441.pdf]. 

National Institute for Occupational Safety and Health (NIOSH), 2004, Report on Residual Radioactive 

and Beryllium Contamination at Atomic Weapons Employer Facilities and Beryllium Vendor 

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